ENS 44168
ENS Event | |
|---|---|
16:30 Apr 25, 2008 | |
| Title | Non-Conservative Core Spray Flow in Dresden Nuclear Power Station Loca Analysis |
| Event Description | Exelon Generation Company, LLC (EGC) has been informed of a change in the vendor calculation of Peak Cladding Temperature (PCT) for Dresden Nuclear Power Station (DNPS) Unit 2. On April 24, 2008, Westinghouse initially notified EGC of a potential non-conservatism with respect to the assumed core spray (CS) flow in the Westinghouse 10 CFR 50, Appendix K LOCA analysis for DNPS Unit 2. Based on further discussions with Westinghouse, EGC confirmed the error on April 25, 2008, at 1130. In selecting the limiting plant configuration for the DNPS Appendix K LOCA model, Westinghouse used the net flow at the CS sparger, instead of the runout flow minus leakage outside the shroud. As a result, Westinghouse used an incorrect input for CS flow for the ECCS LOCA analysis.
Westinghouse has evaluated the impact of correcting the assumed CS flow on the DNPS Unit 2 limiting large break LOCA analysis. Preliminary results indicate an increase of 80 degrees F in the PCT value. This results in a final PCT value of 2230 degrees F, which is in excess of 10 CFR 50.46 (b)(1) limits. However, preliminary calculation results also indicate that a 3 percent reduction in the MAPLGHR limit would offset the 80 degree F increase in the limiting PCT. Dresden Unit 2 has been operating with more than 3 percent margin in MAPLHGR since November 2007 (i.e., the start of the current operating cycle - D2C21). Administrative limits wilt be established to maintain these margins. The LOCA analysis of record for DNPS Unit 3 utilizes the correct CS shroud leakage and CS flow delivered to the vessel through the CS sparger. Therefore, this non-conservatism does not affect DNPS Unit 3. Currently, there is sufficient operating margin for DNPS Unit 2 to ensure the PCT limit would not be exceeded. There are no other 10 CFR 50.46 ECCS criteria for DNPS Unit 2 that are impacted. This notification is being made as a result of the 10 CFR 50.45 (a)(3)(ii) requirement to report this issue in accordance with 10 CFR 50.72 (i.e., 10 CFR 50.72(b)(3)(ii)(B)). The licensee notified the NRC Resident Inspector. |
| Where | |
|---|---|
Dresden  Illinois (NRC Region 3) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | |
| Time - Person (Reporting Time:+1.98 h0.0825 days <br />0.0118 weeks <br />0.00271 months <br />) | |
| Opened: | John Feigl 18:29 Apr 25, 2008 |
| NRC Officer: | Pete Snyder |
| Last Updated: | Apr 25, 2008 |
| 44168 - NRC Website
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Unit 2 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (99 %) |
| After | Power Operation (99 %) |
WEEKMONTHYEARENS 532042018-02-12T18:00:00012 February 2018 18:00:00
[Table view]10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition En Revision Imported Date 10/14/2019 ENS 488452012-11-20T00:00:00020 November 2012 00:00:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Control Rod Drive Check Valve in Service Leak Test Failure ENS 441682008-04-25T16:30:00025 April 2008 16:30:00 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Non-Conservative Core Spray Flow in Dresden Nuclear Power Station Loca Analysis ENS 413702005-02-04T01:15:0004 February 2005 01:15:00 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 4160 Volt Relaying and Metering Single Failure Vulnerability ENS 403112003-11-10T13:20:00010 November 2003 13:20:00 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident, 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition Bypassing the Pressure Suppression Function of the Torus 2018-02-12T18:00:00 | |